The present invention relates generally to nuclear reactors, and, more specifically, to electrochemical corrosion potential sensors.
A nuclear power plant includes a nuclear reactor for heating water to generate steam which is routed to a steam turbine which extracts energy therefrom for powering an electrical generator to produce electrical power. The nuclear reactor is typically in the form of a boiling water reactor having suitable nuclear fuel disposed in a reactor pressure vessel in which water is heated.
The water and steam are carried through various components and piping which are typically formed of stainless steel, with other materials such as alloy 182 weld metal and alloy 600 being used for various components directly inside the reactor pressure vessel.
It has been found that these materials tend to undergo intergranular stress corrosion cracking depending on the chemistry of the material, degree of sensitization, the presence of tensile stress, and the chemistry of the reactor water. By controlling any one or more of these critical factors, it is possible to control the propensity of a material to undergo intergranular stress corrosion cracking.
However, it is known that intergranular stress corrosion cracking may be controlled or mitigated by controlling a single critical parameter called the electrochemical corrosion potential of the material of interest. Thus, considerable efforts have been made in the past decade to measure the electrochemical corrosion potential of the materials of interest during the power operation of the reactor. This, however, is not a trivial task because the electrochemical corrosion potential of the material varies depending on the location of the material in the reactor circuit.
For example, a material in the reactor core region is likely to be more susceptible to irradiation assisted stress corrosion cracking than the same material exposed to an out-of-core region. This is because the material in the core region is exposed to the highly oxidizing species generated by the radiolysis of water by both gamma and neutron radiation under normal water chemistry conditions, in addition to the effect of direct radiation assisted stress corrosion cracking. The oxidizing species increases the electrochemical corrosion potential of the material which in turn increases its propensity to undergo intergranular stress corrosion cracking or irradiation assisted stress corrosion cracking.
Thus, a suppression of the oxidizing species is desirable in controlling intergranular stress corrosion cracking. An effective method of suppressing the oxidizing species coming into contact with the material is to inject hydrogen into the reactor water via the feedwater system so that recombination of the oxidants with hydrogen occurs within the reactor circuit. This results in an overall reduction in the oxidant concentration present in the reactor which in turn mitigates intergranular stress corrosion cracking of the materials, if the oxidant concentration is suppressed to very low levels.
This method is called hydrogen water chemistry and is widely practiced for mitigating intergranular stress corrosion cracking of materials in boiling water reactors. When hydrogen water chemistry is practiced in a boiling water reactor, the electrochemical corrosion potential of the stainless steel material decreases from a positive value generally in the range of 0.050 to 0.200 V(SHE) under normal water chemistry to a value less than -0.230 V (SHE), where SHE stands for the Standard Hydrogen Electrode potential. There is considerable evidence that when the electrochemical corrosion potential is below this negative value, intergranular stress corrosion cracking of stainless steel can be mitigated and the intergranular stress corrosion cracking initiation can be prevented.
Considerable efforts have been made in the past decade to develop reliable electrochemical corrosion potential sensors to be used as reference electrodes which can be used to determine the electrochemical corrosion potential of operating surfaces of components. These sensors have been used in more than a dozen boiling water reactors worldwide, with a high degree of success, which has enabled the determination of the minimum feedwater hydrogen injection rate required to achieve electrochemical corrosion potentials of reactor internal surfaces and piping below the desired negative value.
However, the drawback of these sensors is that they have a limited lifetime in that some have failed after only three months of use while a few have shown evidence of operation for approximately six to nine months.
Recent experience with two boiling water reactors in the United States has shown that the two major modes of failure have been the cracking and corrosive attack in the ceramic-to-metal braze used at the sensing tip, and the dissolution of the sapphire insulating ceramic material used to electrically isolate the sensing tip from the metal conductor cable for platinum and stainless steel type sensors.
The electrochemical corrosion potential sensors may be mounted either directly in the reactor core region for directly monitoring electrochemical corrosion potential of in-core surfaces, or may be mounted outside the reactor core to monitor out-of-core surfaces. However, the typical electrochemical corrosion potential sensor nevertheless experiences a severe operating environment in view of the temperature of the water well exceeding 88.degree. C.; relatively high flowrates of the water up to and exceeding several m/s; and the high nuclear radiation in the core region. This complicates the design of the sensor since suitable materials are required for this hostile environment, and must be suitably configured for providing a watertight assembly for a useful life.
As indicated above, experience with the typical platinum electrochemical corrosion potential sensor has uncovered shortcomings that lead to premature failure before expiration of a typical fuel cycle.
Accordingly, it is desired to improve the design of electrochemical corrosion potential sensors for improving its useful life.